scholarly journals Examination of steam generator tube R45C52 from the Ginna nuclear power plant. Final report

1980 ◽  
Author(s):  
F. Pement ◽  
E. Morgan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Final Report ◽  
Steam Generator Tube

Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack

2016 ◽  
Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Experimental Program ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
Loss Of Coolant ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.

scholarly journals Analysis of steam generator tube rupture accident for OPR 1000 nuclear power plant

2021 ◽  
Vol 382 ◽  
pp. 111403
Author(s):  
Sung Il Kim ◽  
Hyung Seok Kang ◽  
Young Su Na ◽  
Eun Hyun Ryu ◽  
Rae Joon Park ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Steam Generator Tube

Growth Model of Circumferential Outer Diameter Stress Corrosion Crack in Retired Steam Generator Tube of Kori 1 Nuclear Power Plant

2007 ◽  
Vol 124-126 ◽  
pp. 1529-1532
Author(s):  
Dong Jin Kim ◽  
Hong Pyo Kim ◽  
Joung Soo Kim ◽  
Yun Soo Lim ◽  
Seong Sik Hwang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Stress Corrosion ◽  
Steam Generator ◽  
Growth Model ◽  
Nuclear Power ◽  
Forward Direction ◽  
Outer Diameter ◽  
Lateral Direction ◽  
Steam Generator Tube

Growth model of a circumferential outer diameter stress corrosion crack (ODSCC) in a retired steam generator tube of the Kori 1 nuclear power plant was proposed based on extensive destructive examinations of the pulled tubes of Alloy 600 from the Kori 1 plant. A small ODSCC grows in a lateral direction as well as a forward direction until it meets a neighboring ODSCC which also grows in a lateral direction as well as a forward direction. And then, the two ODSCCs which meet on the same circumferential plane are consolidated into a single ODSCC. By repeating such a consolidation process with time, it seems that the apparent growth rate of an ODSCC in the lateral direction is much faster than that in the forward direction. Growth model of a circumferential ODSCC from a retired steam generator tube of the Kori 1 plant reveals that many ODSCCs are initiated and grow in both directions independently until they meet and finally they are consolidated.

Assessment of Limit Loads for Circumferential Cracks in Steam Generator Tube Considering Constraining Effect of Tube Support

2005 ◽  
Author(s):  
Hyun Su Kim ◽  
Jong Sung Kim ◽  
Tae Eun Jin ◽  
Hong Deok Kim ◽  
Han Sub Chung
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Three Dimensional ◽  
Fission Products ◽  
Limit Load ◽  
Steam Generator Tube ◽  
Integrity Assessment ◽  
Steam Generator Tubes

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from reactor to generate steam to drive the turbine generators. Rupture of a steam generator tube can result in release of fission products to environment. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining safety of a nuclear power plant. The steam generator tubes are supported at periodic intervals by support plates and rotations of the tubes are constrained. Although it was reported that the limit load for a circumferential crack was significantly affected by boundary condition of the tube, existing limit load solutions do not include the constraining effect of tube supports. This paper provides detailed limit load solutions for circumferential cracks in steam generator tubes considering the actual boundary conditions to simulate the constraining effect of the tube supports. Such solutions are developed based on three dimensional (3D) finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

scholarly journals Application of Numerical Simulation on the Manufacturing Process Qualification of Nuclear Power Plant Steam Generator Tube-sheet

2016 ◽  
Vol 63 ◽  
pp. 02001
Author(s):  
Shao Jun Zhang ◽  
Shu Hua Liang ◽  
Wen Bin Yang ◽  
Zhao Liu ◽  
Dong Hai Zhao
Keyword(s):  
Numerical Simulation ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Manufacturing Process ◽  
Nuclear Power ◽  
Tube Sheet ◽  
Steam Generator Tube ◽  
Process Qualification

Pulse Electroplating of Ni-P-Nano TiO2 and ZrO2 for Steam Generator Tube Repair

2007 ◽  
Vol 26-28 ◽  
pp. 1067-1070
Author(s):  
Kee Won Urm ◽  
Seon Ho Lee ◽  
Woo Sung Kim ◽  
Chang Yeol Cho ◽  
Jong Ho Lee
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Radioactive Material ◽  
Alloy 600 ◽  
Nano Tio2 ◽  
Steam Generator Tube ◽  
Pulse Electroplating ◽  
Tube Repair

Steam generator tubes provide the pressure boundary between the primary and secondary regions of a nuclear power plant. Alloy 600 is a tube material with good corrosion resistance; however, tubes of this material have experienced damage, particularly as Stress Corrosion Cracking, under the elevated temperature and pressure environment of a nuclear power plant. These damaged tubes must be repaired to prevent leakage of radioactive material from the primary to the second regions in the nuclear steam generator. In this study, Ni-P-Nano TiO2 and ZrO2 layers were produced by pulse electroplating for steam generator tube repair. These electroplate layers were obtained from Ni sulfamate bath with an added small quantity of H3PO3 and Nano TiO2 and ZrO2 particles with an average size of 20-80nm. Results of TEM analysis in these layers show that Nano TiO2 and ZrO2 particles were uniformly distributed into the electroplated Ni matrix and the tensile strength of these layers at 800-1000MPa was higher than that of alloy 600 with a conventional pure Ni electroplate layer.

Leak behavior of SCC degraded steam generator tubings of nuclear power plant

2005 ◽  
Vol 235 (23) ◽  
pp. 2477-2484 ◽  
Author(s):  
Seong Sik Hwang ◽  
Hong Pyo Kim ◽  
Joung Soo Kim ◽  
Kenneth E. Kasza ◽  
Jangyul Park ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power
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